lec18

Operational Reactor Safety

22.091 /22.903

Lecture 18 Pilgrim Nuclear Power Station Background Information

Mark Santiago – Pilgrim Training

1

Welcome to Pilgrim Station and the Chiltonville Training Center

 700 MWe Boiling Water Reactor

 2028 MWth

 Nominal Operating pressure 1030 psig

 145 control rods

 580 fuel bundles

 Mark 1 Containment

 Went commercial in 1972

 License will currently expire in Aug 2012

 Application for 20 Year license renewal submitted

– Expect approval later in 2008.

 Operates on a 24 month refueling cycle

 Owned by Entergy

 Part of a twelve unit nuclear fleet

 Entergy has filed for a permit for new BWR in Mississippi.

Preaa«sañze d Wate v Reacto: r

PRESSURIZER

NUCLEAR HEAT GENERATION

STEAM GENERATION

WATER TRANSFER

STEAM CONDENSATION

TURBINE GENERATOR OPERATI O N

WATER

F¥EACTOR

STEAM GENERATOR

FEED PUMP

REACTOR COOLANT PUMP

Bo:itJog Wate:r Reasto:r

STEAM FLOW TO TURBINE

NUCLEAR HEAD GENERATON

TURBINE GENERATOR OPERATION

WATER SUPPLY FEED PUMP

STEAM CONDENSATION

REACTOR

REClRCULATlOf4 PUMP

CONDENSATE

PUMP

Beacto r an d T u rbin e Building s

ne cTon BuiLDlNc

T UHOINE BUILDING

Reactivity Coefficients

 Void Coefficient

(~ -1X 10 -3 ǻ k/k / % Voids)

 Moderator Temperature Coefficient (~ -1X 10 -4 ǻ k/k / o F)

 Doppler Coefficient

(~ -1X 10 -5 ǻ k/k / o Fuel Temperature)

Critical Power

FORCED CO NVECTIO N HEATING

7

CR ITIC AL POWER

6

2

3

4

5

1

NUCLEATE BOILING

TRANS ITIO N BOILING

FILM BOILING

FUEL ROD SURFACE SUPERHEAT ( T WALL - T SA T )

FUEL ROD SURFACE HEAT FLUX OR

Q ˙ / A

Critical Power Ratio (CPR)

CPR  CP

A P

 1.0

Where:

CPR = critical power ratio

CP = bundle power at which OTB occurs AP = a ctual bundle power

MCPR LIMITS

MARGIN FOR WORST CORE FUEL TRANSIENT

MARGIN FOR UNCERTAINTY IN CORE OPERATING STATE

MARGIN FOR UNCERTAINTY IN ANALYSIS DATA

STEADY STATE CPR

OPERATING LIMIT


MCPR	=	1.06
MCPR	=	1.05

MCPR = 1.0

99.9% OF RODS AVOID OTB

50% OF RODS AVOID OTB


	RADIOACTIVE RELEASE FROM THE PLANT WITHIN LIMITS


FUE L CL ADDING CRACKING DUE TO HI GH ST RES S		GROS S CL A DDING FAI LURE DUE TO LACK OF COOLING	FUE L CL ADDING CRACKING DUE T O LOSS OF COOLING

F U EL PELLE T EX PANSI ON		DECA Y HE A T AND STORED HE AT FOLLO W I NG LO CA	L O SS OF NUCL E A TE BOI LING AROUND C L AD DING

1 % PLAS T I C S T R A IN O N C L AD DING		CL AD TEMPERATURE 2200°F	BOILING TRANS ITIO N

LO CAL F U EL PIN POWER IN NO DE		AV ERA GE FUE L PIN POWER IN NO DE	TOTA L FUE L BUNDL E P O WER

FUL L POWER		LO CA	TRANI ENT OP ERATI ON

LHGR		AP LHGR	CPR

FLPD < 1. 0		MAPRAT < 1. 0	MFLCP R < 1. 0

PURPOSE

FAI LURE MECHANISM

CAUSE OF FAI LURE

LIMI TING CONDITIO N

ITEM MEASURED

LIMI TING OPERAT I O N

LIMI TING PARAMET ER

CALCULATED PARAMET ER

Abnormal Operational Transients

 Abnormal Operating Transients include the events following a single equipment malfunction or a single operator error that is reasonably expected during the course of planned operations.

 Power failures, pump trips, and rod withdrawal errors are typical of the single malfunctions or errors initiating the events in this category.

Reactor Limits

 To avoid the unacceptable safety results for abnormal operational transients, reactor operating limits are specified. Operating limits are specified to maintain adequate margin to the onset of boiling transition and failure due to cladding strain (CPR & LHGR). To ensure that adequate margin is maintained and an unacceptable result is avoided, a design requirement based on a statistical analysis was selected. This requirement would ensure that during an abnormal operational transient, 99.9% of the fuel rods would be expected to avoid boiling transition.

Abnormal Operational Transients Event Categories

 Events Resulting in a Nuclear System Pressure Increase

– Ex: Turbine trip (turbine stop valve closure)

 Events Resulting in a Reactor Vessel Water Temperature Decrease

– Ex: Inadvertent Pump Start

 Events Resulting in a Positive Reactivity Insertion

– Ex: Continuous, inadvertent rod withdrawal

Abnormal Operational Transients Event Categories Cont.

 Events Resulting in a Reactor Vessel Coolant Inventory Decrease

– Ex: Loss of feedwater flow

 Events Resulting in a Core Coolant Flow Decrease

– Ex: Recirculation pump seizure

 Events Resulting in a Core Coolant Flow Increase

– Ex: Recirculation Flow Control Failure - I ncreasing Flow

DESIGN BASIS ACCIDENTS

 A design basis accident is a hypothesized accident; the characteristics and consequences of which are utilized in the design of those systems and components pertinent to the preservation of radioactive materials barriers, and the restriction of radioactive material release from the barriers.

Unacceptable Results

 radioactive material release which results in dose consequences that exceeds the guideline values of 10CFR100

 failure of fuel cladding which would cause changes in core geometry such that core cooling would be inhibited

 nuclear system stress in excess of those allowed for the accident classification by applicable codes

 containment stresses in excess of those allowed for the accident classification by applicable industry codes when containment is required

 overexposure to radiation of st ation personnel in the control room

DESIGN BASIS ACCIDENTS

 Control Rod Drop Accident

 Loss of Coolant Accident

 Main Steam Line Break Accident

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Minus a few hundred thousand gallons of water….

Drywell

 Steel ASME Code pressure vessel

 Shaped like an inverted light bulb

Design Pressure

+ 56 psig

- 2 p s i g

34’ in Diameter 110% overpressure

Vessel

64’ in Diameter

+ 62 psig

Design Temp (Tsat)

281  F.

110’ Tall

Pressure Suppression Chamber

and Pool

Energ y Transfe r Paths

• R elief valve discharge piping

• H P C I

• R C I C

• D rywell Vent system

• R esidual Heat Removal

Hydrogen Event At TMI

 The first warning of the presence of hydrogen in the system was quite violent, but thanks to the heavily over engineered containment structure, it was almost anticlimactic save for its implications. A poorly shielded relay sparked, detonating the hydrogen in the containment. Containment building pressure zoomed to a frightening 28 pounds per square inch, and stayed there for nearly eight seconds as the hydrogen burned. The force shook the control room floor noticeably, and was thought to be equivalent to the explosion of several modern 1,000 pound bombs.

Hydrogen Combustion The Burn

 Deflagrations are combustion waves which heat the gas by thermal conduction

 Travel Subsonically and cause low pressure loads on the containment

Hydrogen Event At TMI Burnt crane controls

Hydrogen Combustion The Boom

 Detonation heats the unburned gas by compression from shock waves.

 The waves travel supersonically and produce high pressure loads on the containment.

Containment Damage

 Hydrogen can create excessive drywell pressure

 Containment design pressure =

 56 psi

 Estimated failure pressure =

 ~ 200 psi

 Estimated pressure with 30% metal-water reaction with a burn

 >> 200psi

Hydrogen Event At TMI ‘Nuf said’

Drywell head

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22.091 Nuclear Reactor Safety

Spring 200 8

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